ANALYSIS OF NEUTRONICS SAFETY PARAMETERS AND CORE BURNUP LIFETIME OF BAEC TRIGA MARK-II RESEARCH REACTOR USING THE DETERMINISTIC TRIGAP AND TRIGLAV CODES

Abstract

BTRR has been operating science 1986 without any form of reloading or shuffling. On September 14, 1986, 50 low enriched uranium (LEU) fuel components were put into the original core, and it reached its first initial criticality. After that, 100 fuel elements were loaded and this configuration is known as the operational core. The goals of this study are (i) to analyze the neutronics core safety parameters of Initial as well as the functioning core of BTRR and (ii) to calculate individual fuel element burn-up as well as ring wise average burnup at different burn-up conditions and core lifetime of the present low enrichment uranium (LEU) core configuration. To analyze the initial criticality experiment of BTRR, its initial critical TRIGA physical model has been developed and hence the initial effective multiplication factor, core excess reactivity has been calculated using deterministic code TRIGLAV and TRIGAP accordingly. Burnup calculations are predicated on the concept that while calculating the neutron density distribution, nuclide concentrations can be taken for granted to be constant. They are built on the neutron transportation calculation and the burnup equations, which are two fundamental equations in reactor physics. Individual fuel burnup as well as ring wise burnup calculation has been done by TRIGLAV code and it has been compared with the MVP-Burn code. Burnup has been calculated up to 1400 MWd and the 1400 MWd data has been compared with the result obtained from TRIGAP code. To estimate the core life time, core excess reactivity has been considered and the calculated results are likened with the experimental obtained values from reactor operational data log book. The reactor may be operated safely for an additional 500 MWd days in accordance with the need of burnup and excess reactivity. This study will be helpful to formulate the most economic use of the fuel rod initially overloaded in the core. Additionally, the study provides insightful information on the behavior of the reactor and will guarantee improved reactor usage and operation in the future. Additionally, utilizing the same fuel components, this can provide insight into redesigning a better core configuration.